Efficient, practical and economically feasible disposal of radioactive wastes is a difficult problem which has commanded consideration attention from nuclear scientists and environmentalists, not to mention the communities and inhabitants of the areas affected by this problem.
Liquid radioactive waste streams are generated in fuel reprocessing and normal nuclear reactor operations. Considering the large quantities of liquid wastes produced in various nuclear installations, efficient cleaning of these wastes and economical disposal of the radioactive cations therefrom are very formidable tasks.
The principal objective in cleaning of radioactive wastes is to separate the radioactive cations from non-radioactive cations and to reduce the waste volume to be disposed. The most common radioactive cations in such waste streams include cesium, strontium, cobalt, barium, yttrium, lanthanum, etc., while the most common non-radioactive cations, or the so-called "inerts", include sodium, lithium and potassium. Also, and in general, the radioactive cations represent a small fraction of the total cations and the inerts represent a small fraction of the liquid waste.
A "high level" waste is produced by dissolving spent nuclear reactor fuel elements. After recovering substantially all of the uranium and plutonium for further use, the remaining solution is usually neutralized with sodium hydroxide for safer storage. This neutralized solution contains large amounts of sodium nitrate salts compared to the radioactive ions present.
The presence of sodium in the solution has several disadvantages. If the sodium-containing waste is stored as a dried salt cake, it is corrosive and will be readily dissolved in water. If the waste is vitrified into a nuclear waste glass, the presence of sodium lowers the chemical durability of the glass and increases the amount of glass which must be disposed.
Thus, in order to effectively clean the liquid nuclear waste, it is important to remove a significant portion, if not all, the sodium ions from the enriched radioactive cation-containing solution. Consequently, ion exchangers are needed which will segregate cesium and strontium from sodium, for example.
Copending application Ser. No. 370,437, filed Apr. 21, 1982, U.S. Pat. No. 4,469,628 which is a continuation of application Ser. No. 39,595, filed May 16, 1979, now abandoned, which is a continuation-in-part of application Ser. No. 959,222, filed Nov. 9, 1978, now abandoned, entitled "FIXATION BY ION EXCHANGE OF TOXIC MATERIALS IN A GLASS MATRIX", describes of a method for the treatment of liquid nuclear waste materials containing radioactive ions by ion exchange with porous silicate glass or silica gel. The porous silicate glass or silica gel employed as the ion exchange media may be prepared in accordance with the methods described in U.S. Pat. No. 4,110,096 to Macedo et al and application Ser. No. 370,437, the disclosures of which are expressly incorporated herein by reference.
In U.S. Pat. No. 4,469,628, nonradioactive cations (i.e., alkali metal, Group Ib metal, and ammonium cations) bonded to silicon through divalent oxygen groups in a porous silicate glass or silica gel matrix are ion exchanged with radioactive ions in the liquid waste. The radioactive ions include radioactive cations and the ion exchange reaction occurs, in particular, on the surfaces of the myriad of interconnecting pores of the silicate glass or silica gel. Thus, the non-radioactive metal or ammonium cations are displaced by the radioactive cations resulting in radioactive cations becoming chemically bonded to silicon or boron through divalent oxygen groups. Thereafter, the resulting porous silicate glass or silica gel may be dried, stored, or packaged or "containerized" in suitable containers or forms, or disposed of as by underground burial or by burial at sea. Desirably, the radioactive silicate glass or silica gel is heated to its sintering temperature to cause partial or complete collapse of the pores and thereby fix and mechanically encapsulate the radioactive cations within the resultant glass matrix.
In concentrating and immobilizing the radioactive cations in accordance with the method described in the aforementioned patent, an essentially "single pass operation" is employed. Once the porous silicate glass or the silica gel is loaded to its capacity, its ion exchange capability is exhausted. This essentially "single pass operation" has limited practical concentration efficiency and economical feasibility where large and voluminous quantities of liquid wastes are involved.
Organic ion exchange media such as Dow HCR-S sold by Dow Chemical Co., Midland, Michigan, have been used to decontaminate radioactive wastes by passing the contaminated radioactive wastes through the organic ion exchange medium. Although organic ion-exchange media are regenerable, the ultimate radioactive loading level of organic ion exchange media is considerably more limited than the inorganic ion exchange media since the former is susceptible to radiation damage at a much lower dosage than the inorganic ion exchange media.
Inorganic ion exchange media have been employed to decontaminate radioactive waste streams. Zeolites, for example, have been used for the extraction of radioactive Cs.sup.+ from containment solution because of their preferential extraction of cesium ions over sodium ions. However, the resultant radioactive-containing zeolites are difficult to strip and regenerate; hence, the concentration of the decontaminated solution is limited to what can be achieved in a single pass operation. Moreover, common inorganic ion exchangers such as zeolites are destroyed by acid. The use of inorganic ion exchange media such as zeolites, therefore, is neither an effective method nor an economical one of treating large volumes of liquid wastes.
Sodium titanates also have been used as inorganic ion exchange media but, in general, they suffer from the same disadvantages which are associated with the use of zeolites.
The use of zeolites and titanates present other disadvantages. The disposal of radioactively loaded zeolites and titanates in a cement mix depends largely on their loading levels. Where the loading level is high, they cannot be disposed of as low level waste and must be vitrified. The loading of zeolites and titanates in borosilicate glasses must be kept low otherwise the glass properties will deteriorate.
There also are problems in separating and purifying non-radioactive cations in various fields such as mining, chemical purification, analytical chemistry and toxic waste treatment.
Accordingly, it is an object of this invention to provide an effective and economically attractive and feasible method of separating and purifying cations from liquids.
It is a further object of this invention to provide a method of decontaminating liquid wastes and concentrating the radioactive cations contained therein by multiple-pass ion exchange with porous silicate glass or silica gel.
It is also an object of this invention to provide a method of purifying liquid wastes involving the removal of radioactive-cations and separating them from the sodium contained therein.
It is still another object of this invention to provide a method of regenerating a porous silicate glass or silica gel ion exchanger to enable multiple-pass ion exchange operation.
It is yet another object of this invention to provide an improved method of selectively stripping cations from a loaded ion exchnage medium.